ML20137B468
| ML20137B468 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 06/30/1974 |
| From: | CONNECTICUT YANKEE ATOMIC POWER CO. |
| To: | |
| Shared Package | |
| ML20137B401 | List: |
| References | |
| FOIA-85-723 NUDOCS 8511260279 | |
| Download: ML20137B468 (58) | |
Text
.
CONNECTICUT YANKEE A'IOMIC POWER COMPANY HADDAM NECK PLANT HADDAM, CONNECTICUT DOCKET F0 50-213 i
e J
SENIANNUAL OPERATING HEPORT NO. 714-1 j
FOR THE PERIOD OF i
JANUARY 1, 19714 to JUNE 30, 19714 August 20, 19714 l
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TABLE OF CONTENTS Page Operations Summary.
1-5 Plant Shutdowns
.6-7 Corrective Maintenance 8 - 18 Plant Changes, Tests and Experiments
. 19 - 23 Major Changes in Facility Design in Progress
. 23 License Application Amendment
.2h Technical Specification Changes 2h - 25 Plant Operating Procedures
. 26 - 28 Abnormal Occurrences 29 - 37 Unusual Events 38 Changes in Key Personnel.
38 Power Generation Summary.
39 Histograms (Thermal Posar/ Time)
LO - h5 Plant Performance (Primary).......
. h6 - 48 Plant Performance (Secondary) 48 - h9 Surveillance 49 - 51 Primary Coolant Chemistry.
52 Solid Waste Shipments 53 Transfer of Special Nuclear Materials 53 Occupational Personnel Radiation Exposure 5h Radioactive Effluent Releases (Liquid) 55 Radioactive Effluent Releases (Gaseous) 56 Environmental Monitoring.
57 e..
h
i OPERATIONS
SUMMARY
The following is a chronological description of plant operations for the six months report period ending June 30, 1974.
1-01 Load increase in progress.
1-02 Full load (600 MWe) attained @ 0328.
1-13 Load reduced to 400 MWe for routine turbine stop and control valve testing.
1-lh Full load (597 MWe) attained @ 0555 1-18 Automatic plant trip @ 0613 from 601 MWe.
1-18 Generator phased to grid @ 1255 Increased 1 cad reaching h80 MWe (80% power) @l903 1-19 Automatic plant trip @ 0406 from 480 MWe.
1-19 Generator phased to grid @ 0929 1-19 Automatic plant trip @ 12kT from 300 MWe.
1-19 Generator phased to grid @ 1657.
Increased load reaching I
h80 MWe (80% power) @ 2207 1-20 Full load (600 MWe) attained @ 2350.
1-25 Temporary Primary Auxiliary Building Ventilation Exhaust System was placed in operation.
1-26 Load reduced to 400 MWe for routine turbine stop and control valve testing.
1-26 The 1250 Line and 399 Station Service Transformer were removed from service at 0511 for protective relay maintenance.
1-27 The 1250 Line and 399 Station Service Transformer were returned to service @ 0033 1-27 Full load (601 MWe) was attained @ 1203.
2-02 Load reduction to 513 MWe and subsequent return to 100% power during transient caused by closure of #2 steam generator feedwater valve.
2-10 Load reduction to 400 ! awe for turbine stop and control valve testing and subsequent return to 95% power.
M M
t@p W
M
2-11 Load increase to 100% power following 2h hour hold at 95%.
2-15 Load reduction to 300 MWe for repairs to #- steam generator feedwater valve.
2-16 Turbine stop and control valve test while @ 300 MWe, 2-16 When repairs to kh feedwater flow control valve were
]
not able to be done on-line, the secondary plant was l
shut down, the generator being ceparated from the grid @ 0657 2-16 The generator was phased to the grid at 1831 and load increase to 80% commenced.
2-17 Load reached 80% power @ 0230.
2-18 Load increase to full power completed at Oh00 following 2h hour hold @ 80%.
2-22 Load reduction to h00 MWe for turbine stop and control valve test and condenser tube plugging.
2-23 Load increase to 480 MWe (80%) for 2h hour hold.
2-24 Load increase to 100% following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> hold at 80%.
2-24 1772 line was removed from service for line work between 0850 and 1231.
3-9 Commenced load reduction to 400 MWe @ 0450 to conduct a turbine stop and control valve test, locate and plug leaking condenser tubes and data acquisition for core axial offset study.
400 MWe load reached @ 0620. A satisfactory valve test was conducted and 28 condenser tubes plugged.
3-9 Commenced load increase @ 1930.
Reaching h80 MWe (80% power)
@ 2022.
3-10 Completed 2h hour holding period and commenced load increase @ 2103. Full power (59h MWe) was reached 6 232'0.
3-12 Circulating water intake warming line was removed from service.
3-21 An increase in turbine vibration readings occurred @ 2120.
Maximum recorded was 7.6 mills (still less than the maximum allowable operating limit.)
e 2-h 12
3-23 Commenced lond reduction to 400 MWe @ 0600 to conduct a turbine stop and control valve test, locate and plug lenking condenser tuben and to further evalunte increased turbine vibration readings.
h00 MWe lond was reached
@ 0800. The turbine valve test was terminated because of problems with #1 control valve test motor (motor burned out). Lond reduction commenced @ 1140 for mensurement of turbine vibrntions nt minimum lond.
80 MWe vns reached @ 13h8. Analysis of the turbine vibration data indiented a need for an internal inapection so load was reduced and the generator separated frcm the grid @ 1617. Ten (10) condenser tubec vere plugged (3 in "A" and 7 in "C") during the reduced lond period.
3-23 Turbine disassembly commenced.
3-2h Generator purged of !!.
2 3-25 Turbine-Generator off turning gear.
The reactor vns shutdown @ 1202.
It was returned to criticality @ 2333 for training and then chutdown ngnin
@ 2335 Cooldown of the primary pinnt commenced to perform maintenance on. the primary system during the turbine outage.
3-26 Residual heat removal system was pinced in operation
@ 1858.
3-27 Reactor coolant cyotem vns de-gnaned and cooled down to 127 F by 06h5. The pressurizer relief tank vna depressurized and purged, the pressurizer vented and at atmospheric preneure by 0700. Residual hent removal purification cystem vac pinced in cpuration 81800.
3-28 The plant was in n cold nhutdown condition, both low precoure turbine rotom removed for inupection and repair, generator nnd exciter open for innpection, utenn generntors in vet lay-up, mincellaneous maintennnee on both primary and accondary pinnte in progrena.
h-1 The pinnt was in n cold nhutdown condition with the Residuni llent Removal System in operntion nnd mincellnneoun pinnt maintennnce in progrena.
h-7 Renctor Coolnnt Cyntem filled, vented, hented up to 22h F, preocurized to 2000 psig for n leakage check, llentup to operating tempernture commenced at 2259.
3.
4-8 Bubble vna drawn in the pressurizer @ 0500 and the hentup of the Reactor Coolant System to operating temperature completed @ 1010.
4-9 Hequired nucient instrumentation and miscellaneous surveillance items completed.
f h-10 Beneter achieved critic #tlity @ 0314 l
h-10 to L-17 Penetor pinnt wnu maintained 11. n hot standby condition and the reactor used for training.
h-18 Turbine vna placed on turning gertr @ 0935. Turbine was rolled @ 2145, brought up to synchronoua speed.
then ohut down for balnneing.
h-19 Turbine balnneing continued.
L-20 Generator phased to grid 0 0423.
Lond increnned to
=30 MWe then shutdown @ 05h6 for balttncing. Several more bninneing moyen vore completed during the day with the generator being phnned 8 0953, 1458, 2004 and chutdown @ 1115, 1547 and 2008.
4-21 Generator van phnned @ 0356, removed from the line 6 0435, balance move completed and phased ngnin
@ 1028. This completed the turbine balancing progrrim ao lond van increnced in nteps to 330 MWe by 2005 i
l h-22 Lond inerenced to h00 MWe by 0615, then placed on 2h hour hold.
h-23 Load inerenue commenced 0 0650 renching full lond (596 MWe) @ 0830.
l 5k At 1100 m lond reduction to 400 MWe commenced for et turbine vrtive tout rtnd condenaar tube plugging. The i
pinnt wrta returned to 95% power nt 1636.
5-5 The pittnt vna returned to full lond ttt 1000.
5-9 bl60V-480V ntrttion nervice trannformer #h85 frtiled.
Power for 4807 Dun #5, normally red by thin transformer vna in.acdintely rentored by interconnecting with 480V d
Dun #h.
5-11 At 1207 a lond reduction to 400 MWe commenced for a turbine visive tent, then subnequently reduced to 300 MWe for maintennnec work on ID ntenn generntor reed vnter pump / motor rtilgnment check.
A.
A
5-12 A temporttry 14160V h80V trnnaformer vna instralled nnd reed to h80V Dus #5 restored to normal.
5-13 Iond uno restored to 100% 0 0150.
5-25 At 0700 a load reduction to hoo MWe commenced for a turbine valve test, then cubacquently reduced to 300 MWe for further investigative work with 1D aterun generntor feed water pump vibration problem. Work wto completed at 1708 ttnd lond increased, reaching 80% rull power
@ 1911.
5-26 Following 80% power 2h hour hold, lond uns inerenced reaching 100% nt 21I5 4
6-8 At 0700 a lond reduction to h00 MWe commenced for routine turbine stop and control valve teatr..
The plant van returned to 95% power nw 1100, 6-9 The plant wnn returned to full lond at th10.
6-14 At 2110 n lond reduction to h00 MWe commenced for routine turbine valve tenting.
Load una further reduced to 300 MWe R 0040 for maintenance rtnd innpection of the IB etertm generator reed pump. After neveral balnneing moves and bearing innpection, the IB reed punp vrto returned to service.
While nt reduced lond, rtil condenaer water boxes were inopoeted and condenner lonk checka vere perforned. One tubo van plugged in "D" condenser.
6-15 tond inerence cor:menced at 1926, ree.ching 80% power at 2138 for rs required 2h hour hold.
6
- ,7 I,ond vno rentored to 100% @ 0015.
6-P2 At,1955 there van rtn unexplrtined 15 MWe lond decrence.
l The Control Cynten wnn innpected and van round normal.
Lorid van rentored to 100%.
6-?h At 1003 nn nutomatic renctor/ turbine trip occurred from rull power. The trip wnn enuned by rt momentary lonn or renetor coolttnt riov nignrtl from Loop 2.
The flow trnnnmitter for Inop 2 wnn repiriced, tented nnd returned to nervice. The ronctor wnn returned to crittenlity nt 1800. At 1847 th" #pnerator van phnned to the electrient grid rind londing commenced.
6-P5 fond renched 80% nt 0327 nnd was held at that, level for the required 2h hourn.
6-26 At 0730 lond wnn rentored to 100% (575 MWe).
6-30 Full lond at end or report period.
5-
P' A::: S*1'DO *.':S TYPE METHOD OF F-FO?CID DURATION SHU?OINC DC'43 P:.A*IT S~A';US CO!?!I'ITS/CAUSE OF OUTAGE /
!!O.
DATE S-ScicOULD fu'roS)
PEAS ^*I (1)
THE PEA Z R (?) DURI!!G COTAGE COFFECTIVE AC" ION i
i 105-5 L'TL-01-lS F
6.7 A
Auto =atic orip ' Hot standby During extrenely cold weather cond-AO TL-2 itions, freezing of sensing lines to
- 1 and #2 steam line excessive ficv i
transmitters produced a sinulated i
high flev condition, satisfying the j
i i necessary trip 1cgic (2/L), causing l a reactor / turbine trip. The lines
{
i
' vere thaver'., instrunentation returned l
to normal. Additional area heating g
capacity was provided and several openings closed to prevent a re-currence. Non-nuclear.
i
& 106-5-5 fL-01-19
?
5.33 l
A Autccatic trip Hot standby During a se'.ere ice stern a =cnentary a
8 AO TL-3 fault on the 1572 line coupled with g
i incorrect protective relaying action caused a loss of both cff-site power i
supplies (total loss of AC) resulting l
in a reactor / turbine trip. The l
prctective relays were adjusted to 8
proper settings to prevent a re-(
u eurrence. Non-nuclear.
l
.107-5-6 TL-01-19 F
L lT G
Autccatic trip Hot standby The operator inadvertently shut down j
two circulating water pu=ps supplying the same condenser, satisfying the necessary trip 1cgic and causing a reactor / turbine trip. Ncn-nuclear.
10S-5-T TL-02-16 F
11.6 3
Not Hot standby Secendary plant shutdown to repair applicable (Beactor leaking flange on #L feedvater control critical) valve. *!on-nuclear.
I I
r I
8
s l
I I
TYPE METHOD OF
~
F-FORCED DURATION SHUTTING DOWN PLANT STATUS CO.44ENTS/CAUSE OF OUTAGE /
No.
DA"'r S-SCHEDUTrn (FUPS)
REASON (1)
THE REACTOR (2) DURING OUTAGE CORRECTIVE ACTION 109-5-8 Th-03-23 F
660.10 A&B A
Hot standby Investigation of increased turbine to cold shutdown vibration resulted in a plant shutdown to conduct an internal in-
'l spection of both #1 and #2 low pres-sure turbines.'
'(disegsed
+
I J21 blade;,[n
..s aEeL r50 7 ' )
l GoveNobndm,9f.;..i.kPf$.tsrbine[^ ~ ^
I l
pr.okenlofLaboyeJ,he root. platform.
The 'ishroud,was.r.issing from.the '
entiregrouhY(blade #23+#28).
Non-nuclear.
i-n0-5-9 7k-Ok-20 F
h.n B
Not applicable Hot standby For balance move on turbine.
Non-nuclear.
. nl-5-107h-Oh-20 F
3.72 B
Not applicable Hot standby For balance move on turbine.
Non-nuclear.
Y na-5-nth-Ok-20 F
h.29 B
Not applicable Hot standby For balance move on turbine.
Non-nuclear.
n3-5-12Th-Ok-20 F
7.80 B
Not applicable Hot standby For balance move on turbine, i
Ncn-nuclear.
t nh-5-13 7h-Oh-20 F
5.89 3
Not applicable Hot standby For balance move on turbine.
[
l Non-nuclear.
I
.n5-5-lh Ik-06-2h F
S.73 A
C Hot standby Defective filter capacitor in the amplifier module to Loop #2 flov transmitter intiated reactor trip signal causing unit trip. Amplifier l
module was replaced, channel re-l calibrated and setpoints checked.
l (Nuclear)
(1) REASON:
(2) METHOD:
A-EQUIPMENT FAILURE (EXPIAIN)
E-OPERATOR TRAINING AND A-MANUAL
?
B-MAINTENANCE OR TEST LICENSE EXAMINATION B-ENUAL SCRAM l
C-REFUELING F-ADMINISTRATIVE C-AUTOMATIC SCRAM r
l D-REGUIATORY RESTRICTION G-OPEPATIONAL ERROR (EXPLAIN) l
,i a
.,e
,_---n--
a n - -,.
i-EFFECT SP;TT A'. P.1CC/.'JTION5 SYSTEM ON CORRECTIVE ACTION TAXEN TO PROVIDE, OR MALFUNCTION SAFE TAKEN TO PREVENT FOR REACTG2 SAFETY COMPONENT CAUSE I
RESULT OPERATION REYETITION DURING REPAIR
\\1 Pressurizer Heaters Breaker "E" Group of None Replaced defective shunt trip Not applicable Backup Group "E" nalfunction backup pressu-coil in breaker rizer heaters unavailable e.
Refueling Water Tube leakage Leakage of None Plugged two leaking tubes.
Not applicable Storage Tank Thermo-radioactive, syphon Heater borated liquid into the heating condensat e return system Demineralized Water Head gasket Water leakage None Replaced head gasket.
Not applicable Storage Tank Thermo-leaking to heater pit syphon Heater EG-2A Starting Air
. Cylinder head Reduced air None Replaced cylinder head gasket.
Not applicable.
Co= pressor gasket leakage makeup supply for EG-2A starting air system Boron Recovery Mechanical seal Contaminated None Installed new bearings and Not applicable Syste= Recycle Punp leakage and floor area techanical seal, failed bearings Waste Disposal Leaking valve Slightly con-None Replaced teflon "0" ring seals Not applicable.
System Valves body seals taninated floor WD-V-132 and area j UD-V-lhi Spent Fuel Pcol Packing gland Cont aminated None Repacked valve gland.
Not applicable Recirculation Systen leakage floor area Valve SF-V-813 Flux Mapping Syste=
Faulty Spikes in flux None Replaced faulty detector.
Not applicable.
"B" Flux Detecter detector traces
}
r
, _ ~
96 31,..?:3
, ci CORECTIV." AC1 C.,
i TM'.
J2..~
j S; U TAKEN 10 PR2VE:"!
F03 k.l.C!;!. ' A 3
MALTLTOTION lGPERATIONl RE?ETITIOX OC C T REPAIR
^'3 t.lo :SE i
RESULT I
Refueling Water Dirty regulator. Erroneous jNone Cleaned regulator.
I Visually verified Storage Tank Level reading l
norn.a1 vater level in L
Transmitter Air 8
i e
RWST.
Supply Regulator l
',NIS Channel 21
, Loose connector Low reading lNone Tightened connector.
Maintained Intermediate Range at teter l
I surveillance of other UIS channels I
ff.S Recorder Loose Erratic None l Tightened loose connection.
Not applicable connection readings NIS Channel 32 Faulty card Spurious over-None Replaced faulty Q1 transistor.
Maintained manual Power Range power trip control of signal pressurizer pressure.
1 Pressurizer Level Faulty light Core cooling None Replaced light source Not applicable Isdicator LIA 401-3 source alarm I50 8 Flux =apping syste=
Faulty micro-Withdrawal Mone Replaced micro-switch Not applicable switch interlock in-operative 02 Stea Generator Faulty Relay Caused feed-None Replaced alarm unit
- 2 steam generator High Level Over-water ficv feedvcter control ride Relay control valve placed in manual to close Loop #2 AT Channel Poor lead a? indication None Installed 1 962 ch: resister Trip calculator resistance hi6her than in Tc leg signal fro Loop #2 actual placed in defeat Loop #2 Tavg Calibration Tavg indication None Recalibrated channel Trip calculator Channel drift 1 4 higher than signal fron Loop 72 actual placed in defeat Loop WL AT Channel Loose Indication none Tightened loose connection and Trip calculator connection lover than l
checked calibration signal from Loop *L actual placed in defeat l
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COERECTIVE eCTION TAF.C; TG iG1/ILE S/.FC I
TAKEN TO P2F. VENT FOR 2 J.C!CR SAFETY j
"ALFUNCT ICN v1 jOPZRA: ION REJETITION DURING REPA11 CCy?ONENT t
CAUSE RESULT 5
l I
RMS Co=ponent Coolin6 l Rear connector Lov alarm None Repaired rear connector.
Not applicable l
Lov reading None Repaired indicator.
Used alternate tenp.
lFaultyindicator
- 3 RCP Lcwer Searind 1
Temperature indicator Pressurizer Level Signal failure Alarms, poor None Cleaned gain switch.
Manual control by Control auto, control Operations NIS IR Channel 21 FR drawer Insensitive None Replaced PR drawer.
Not applicable
'NIS Flux Recorder T16ht clutch Paper hard to None Repaired chart roller clutch.
Not applicable turn RMS Recorder Inder svitch Points 1 & 3 None Cleaned index switch.
Not applicable l
not printed I
I RMS River Effluent Meter connector !Hi alarm None Repaired connector.
Not applicable i
,RMS Cc=p. Cooling Erratic rate =eter. Hi alarm and Hone Repaired ratemeter.
Not applicable lov readings r.
, un I
9 l
Feed ater System Capacitors Setpoint drift None Replaced affected capacitors in Not applicable l
all trip and alarm relays (SIGMA) l l
RCP Lev Flev Orip Relay Failed relay Channel 1 & 2 None Replaced failed relay.
Not applicable trip signal i
- 1 Charecal Spray Age Leakage None Replaced head gasket.
Not applicable Valve Strainer Spent Fuel Pit Nor=al wear Leakage into None Replaced pump bearings and Not applicable Cooling Pu=p S.F.B. floor nechanical seal.
FA5 Air Moniter Open coil Motor stoppage None Replaced motor.
Not applicable EG23 l= proper circuitry Overheated EG-2B Removed neutral lead frcs "Y" EG-2A available and design
- excitation trans-unavailable connected diesel den. excitation tested / Tech. Specs.
jfor=erleading (AO Th-8) transformer.
lto failure l
o 4
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T.2.K EN TO PREVLhT
' FOR ?Lt.CTC". :;7;rY PATIONCTION SA?d CCM?JNU;I CAUSZ l
RESULT 0? ERA'f0; JEFETITION DtT.7"r R-P/GR i
i i
EG2A Improper High amperage None Removed neutral lead from "Y" EG-2B available icircuitry design 'on excitation connected diesel gen, excitation transformers transformer.
- k Charcoal Spray Improper fit Seat leakage None Corrected seating angle of the Not applicable Velva lbetweendiscand disc.
Isaat i
Charging Pu=p Errosion Valve body leak None Replaced valve.
Not applicable Mini===
R: circulation Orifice Bypasa Valve EG2A Lube Oil Nor=al wear Noisy motor None Replaced motor.
Not applicable Circulator Pu=p bearings Station Service Unknown Internal fault None Installed temporary transformer k807 Bus 5 inter-hl60-h80V of same rating connected and Transfor:er #L85 l
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energized through h807 Bus b
- cs
's SG"P Recire. AGV Flexible line Valve opened None Replaced air lines on both recirc. Not applicable Plnstic Air Line broken valves with cloth impregnated rubber hose.
- 3 Cent. Recire. Fan Linkage out of Slow cperation None Adjusted linkage.
3 other fans available Face Da=pers adjusteent Reg. Ex Chg. Line Failed data Poor indication None Peplaced failed light source.
Not applicable
- Te=p.
logger signal Pressurizer Press.
Dirty slidevire Indication poor.
None Cleaned slidevire.
Manual control of Pzr.
Pse.
Control sluggish htrs. and spray.
in auto.
Pressurizer Relief Light source Poor indication None Replaced light source in data Not applicable Tank Te=p. Ind, logger repeater.
1
- 3 R0P Seal '='ater Defective Calibration
!Ncne Beplaced capacitors.
Checked local ind Return e_ lev Ind.
capacitors error I
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CO.GECYiVE 3.CTION TAK"N I) I'RC'!IDE pa,.
g ux MAI.FU';CT ION SA7i TAKE" FC PREVENT FCR RE/.CTOR SAFETY COMPONi3T CAUSE i
RESULT CPEASTION l R Z.0ETITION DURINC REPAIR I
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B Charging Funp Lack of bearing lHigh bearing None Replaced inboard bearing, checked Not applicable Motor llubricatien due temperature alignment and cleaned burrs off to stuck slinger slinger.
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C2 Main Steam Line Dried out
? Valve did not See Re-adjusted packing gland Not applicable Trip Valve
! packing operate during AO 74-10 followed by success ful test.
surveillance test s
l "D" Main Condenser Unknown
' Leaking tube None Plugged tube.
Not applicable
'#3 variable Lev Normal Trip, setpoint See Replaced the trip unit and Trip signal inserted Pressure Trip Unit instru=ent drift found outside A0 TL-11 adjusted setpoint to allev for to satisfy the 1/2 Tech. Specs.
normal instrunentation drift.
logic required for L
limit li continued full power l
operation.
Nuclear Overpower
}Nor=alinstru=ent Trip, setpoints See Adjusted the setpoints to allow Not applicable Trip Units
'iri ft found outside Ao TL-9 for nor=al drift without Tech. Specs.
exceeding applicable limits.
limit R: actor Coolant Faulty transmitte Mo=entary loss None Replaced transmitter anplifier Not applicable Loop #2 Flev a=plifier unit of flev signal unit.
Trans=itter resulting in nlant trip 1
Steam Generater Slipped sti False alar =
None Replaced shi= and restored alar =
Not applicable l
Eold Levn Bolt alleving sensing g system to nor=al.
Monitoring Syste:
- switch *o alar =
l A E 3 Centai=nent
' Coupling vear Vibration and None Replaced couplings and realigned Not applicable Service '.'ater leakage sane.
Filters (Ada=s) i j
Rcsin Casi Nor=al wear Pu=p inoperative None Replaced pu=p inpellor.
Not applicable Devstering Pu=p i
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CORRECTIVE ACTION W.C. ' I' ~' T.
0 i-Ox MALFUNCTION SA/C TAKEN
'JO PREVENT FOR LEr.t.OR SAF:iY
- i COMPC::ENT l
CAUS2 i
RESULT OPERATION REPEr1 TION OL?::!' REPr.IR 1
i Diesal Fire Pump
-l Normal wear Exhaust fumes None Replaced expansion joint.
Not applicable
- Exhaust Expansion in screenhouse Joint l' Dicsel Fire Pump Sediment in oil Pump wouldn't None Cleaned solenoids, and Not applicable and cooling start when hot thermometer well and replaced systems and gave false thermometer.
temperature indication
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"A" Control Air Failed head Air leakage None Replaced head gasket.
Not applicable Compressor gasket
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PI ANT CHANGES, TESTS AND EXPERIMENTS The following is a brief description and summary of the safety
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evaluation for those changes which were carried out without prior commission approval. Pursuant to the requirements of 10 CFR Part 50, Section 50.59(b).
Plant Design Change #139 Plant Design Change #139 entitled, " Motor Operated Disconnect SV 320-12R-5" vas completed.
This change concerns the installation of a
" Kirk" interlock between 320-12R-5 (Generator Disconnect) and 12R-lX-9 (Generator Ground Blade) and the installation of a blocking relay in the motor operated disconnect switch control circuit to prevent inadvertent operation if parts of the circuit become grounded. This change will improve the reliability and safety of operations of the disconnect switch and ground blade.
It will also up-date the system to present standards. The safety evaluation considered this change to be non-safety related and does not constitute an unreviewed safety item.
Plant Design Change #1h3 Plant Design Change #1h3 entitled," Service Water to North Side of Plant" was completed. This change concerns the installation of a line from the drain connection on F1-53-1A, Primary Plant Service Water Filter, q
to the north side of the primary auxiliary building 8 23'. This line will provide a temporary supply of service water for construction purposes and a permanent supply for cooling air compressors used during containment pressure testing. The safety evaluation concluded that:
- 1) The probability of occurrence and the consequences of an accident or equipment malfunction are not increased by this plant change.
- 2) The possibility for a different type of accident other than that previously evaluated does not exist.
- 3) The margin of safety is not reduced.
h) Tne change represents sound.en6 neering judgement.
i Plant Design Change #163 Plant Design Change #163 entitled," Temporary Ventilation and Purge System" was completed. This change concerns the installation of a temporary 35000 SCFM fan on the roof of the primary auxiliary building with temporary connections to the PAB ventilation syctem and the containment purge system.
Existing flow instrumentation will also be temporarily relocated from the existing to the temporary system. This change vill provide the necessary ventilation and purging capability needed for the plant while the existing system is being rebuilt.
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The safety evaluation for this temporary plant design change has been reviewed. The installation of a temporary roof fan vill be capable of providing dilution air for possible hydrogen and radioactive gas leakage j
in the air during normal operating and maintenance periods. The containment building purge system may be operated using this temporary fan while isolating the PAB ventilation requirements.
The probability of occurrence and the consequences of an accident or equipment malfunction are not increased by this plant design change.
The possibility for a different type of accident other than that previously evaluated does not exist. The margin of safety is not reduced. This plant design change represents good engineering judgement and vill allow the replacement of the existing ventilation and purge fans and ducting.
Plant Design Change #166 Plant Design Change #166 entitled," Cable Vault Heating System Modification" was completed. This change concerns changing the present cable vault heating system heat supply from steam coils to electric, including the removal of steam and condensate lines from the cable vault.
This change vill remove the possible environmental effects on safety related electrical cables from a break or rupture in the steam and condensate lines located in the cable vault and also provide for area heating, as stated in a report to the AEC on June 29, 1973 This plant design change has been evaluated for safety with the following results:
1)
The probability of occurrence and the consequences of an accident or equipment malfunction are not increased by this plant design change. This change eliminates the possibility of a high energy pipe break inside the cable vault area.
2)
The possibility for a different type of accident other than that previously evaluated does not exist. The electric
)
heater is connected to a non-vital AC bus.
3)
The margin of safety is not reduced, b)
The change represents sound engineering judgement.
Plant Design Change #1hD Plant Design Change #1h0 entitled, " Fuel Oil Transfer Pump Priming System" was completed. This change concerns the installation of a 2 1/2' section of h
ID pipe in the pump discharge line with a 1/2" tie line from the bottom of the h" pipe to the pump suction priming connection. A solenoid operated valve vill be installed to open when the pump motor is energized and close after a time delay. The reason for this design change to the fuel oil transfer pumps fo'r EG-2A and EG-2B fuel oil systems is that periodically the pumps vould lose their prime between operating cycles and the operator vould _have to manually prime the pumps. This change vill ensure the
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automatic priming of the pump suctions.
The safety evaluation for this change is:
1)
The probability of occurrence and the consequences of an accident or equipment malfunction are not increased by this plant design change.
2)
The possibility for a different type of accident other than that previously evaluated does not exist.
3)
The margin of safety is not reduced.
h)
The change represents sound engineering judgement.
Plant Design Change #156 Safe-Guards Equipment"$han g #,156,enti,tled,c" Flooding Protection of PlaritlDesign was completed. This change concerns the installation of the following equipment and components:
Liquid Level Detectors:
In order to alert the operator te a flooding condition as soon as possible, redundant liquid level detectors vill be installed at lh key locations, in the primary auxiliary building, the two emergency generator rooms, and the screen well house. A new panel vill be installed on the main con,rol board and will contain visual and audible alarms from each detector.
Dikes: Carbon steel dikes vill be installed at four locations at elevation 21 ft. - 6 in, and one location at elevation 35 ft. - 6 in.
of the primary auxiliary building to protect charging pumps, metering pump, residual heat removal pumps, low pressure safety injection pumps and high pressure safety injection pumps.
Barriers at Floor Openingsi, Carbon steel barriers vill be installed around floor openings at ele.ation 35 ft. - 6 in. in the primary auxiliary building which are above safeguards pumps. They vill alsc be installed at elevation 21 ft. - 6 in. around floor openings above the Residual Heat Removal Cubicle.
Door Louvers:
In order to supply a path for water to leave the primary auxiliary building, three new doors will be modified and equipped with louvers that vill open under a few inches of water pressure.
' Expansion Joint Sleeves:, The rubber expansion joints in the dis-charge piping of the circulating water pumps offer a possible, although highly unlikely, point. of failure.
In order to prevent-flooding if failure should occur, carbon steel reinforced rubber sleeves will be installed to completely encase the existing expansion joints. ~
The reason for the change is to protect the service water pump motors, chEEgiliU~tmip*h6t'6^5i,~~m'e't'eri'n~g' pump motor residual heat removal' MMotod,ghpres,stire safdt/Tn'jeEt'idn*pudp'r,io' tori"n' d 'emerdency7 diesel
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, generators-from -flooding.
This change will complete our commitments to the AEC " Flooding of Critical Equipment" letteridated August 8,1972 and provide protection for the above safeguards equipment.
The safety evaluation for this change is as follows:
1)
The probability of occurrence and the consequences of an accident or equipment malfunction are not increased by this plant design change.
2)
The possibility for a different type of accident other than that previously evaluated does not exist.
3)
The margin of safety is not reduced.
b)
The change represents sound engineering judgement and will increase plant reliability.
Plant Design Chance #131 Plant Design Change #131 entitled, " Gland Heating Steam Supply Valve Closure" was completed. This change concerns the modification of the high pressure turbine gland heating steam supply valves control circuit to provide valve closure (by de-energizing solenoids) on a turbine trip.
Terminating steam supply to tais system on a turbine trip will eliminate a potential contribution to turbine overspeed following a loss of load condition.
This change was evaluated against criteria stated in 10 CFR 50.59 and does not constitute an unreviewed safety question. Basically it only involves a change to an auxiliary system for the turbine with the end result being the improved operation of the unit.
i Plant Design Change #153 Plant Design Change #153 entitled, " Reactor Coolant Low Flow Trip Matrix" was completed in the month of April, 1974. This change involves the addition of a terminal strip through which all connections to the
" Couch" relays forming the trip matrix vill be routed. At present, all connections to these relays are soldered and cannot be de-energized.
If a relay failed while on line it could not be removed and replaced without a high risk of a plant trip. The change will provide a means of isolating a failed relay and eliminate any possibility of an accidental plant trip during replacement.
l This design change is safety related, but it does not constitute an unreviewed safety item since it does not increase the probability of occurrence of the consequences of an accident or malfunction of equipment
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important to safety previously evaluated in the safety analysis report, and ft does not reduce the margin of safety ns defined in the Technical Specification.
Plant Design Change #1h7 Plant Design Change #1h7 entitled, " Containment Cooler Air Inlet and Outlet Temperature Indication" was completed in the month of May,1974.
This change involved the installation of local temperature indicators on the inlet and outlet plenums of the four containment air recirculation fan / cooler units. Up to this time, there was no installed capability of measuring differential air temperatures across the coolers for performance determination. The addition of these temperature indicators will provide this capability.
Although these indicators are installed in engineered safeguards equipment, the installation does not constitute an unreviewed safety item.
It does not increase the probability of occurrence or of the consequences of an accident or malfunction of equipment important to safety previonely evaluated in the Safety Analysis report, and it does not reduce the margin of safety as defined in the Technical Specifications. The change vill help provide better results for cooler performance surveillance.
Plant Design Change #170 Plant Design Change #170 entitled, " Liquid Overflow and Vent Modification of Liquid Storage Tanks" was completed in the month of June, 1974.
This change involved the following modifications to the Refueling Water Storage Tank: 1) Extend the existing roof vent to overflow into the existing boron vaste storage tank dike sump area; 2) Seal the roof hatch cover with a gasket; 3) Insulate and heat trace the vent line. The two
. test tanks were modified by uncapping the roof vent and extending the vent to overflow into the existing boron vaste storage tank sump area.
The safety evaluation concluded that the tanks in question can be vented to the atmosphere because they are aerated and no possibility of uncontrolled gaseous release exists. This change vill not in any way offset the Category I function and integrity of the Refueling Water Storage Tank.
MATOR CHANGES IN FACILITY DESIGN IN PROGRESS Major changes in facility design in progress during the report period are changes to the radioactive-vaste systems, as reported in the previous semi-annual operations report. The upgraded gaseous vaste system vill provide a means of removing dissolved gases from plant liquids, compressing and storing them. The aerated liquid vaste system vill provide filtration, ion exchange, holdup, evaporation and drumming of contaminated, aerated liquids. The existing primary auxiliary building ventilation and containment purge systems is being modified to include prefilters and charcoal filters. The entire upgraded system is expected to be operational by the fall of this year. The original schedule has been delayed by poor deliveries and changes in system design.
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l LICENSE APPLICATION AMENDMENTS i
Amendment No. 2h to the license application dated September 6, 1963 (Docket No. 50-213) was submitted to the Commission effective January 7, 1974 This amendment is entitled, " Industrial Security Plan" and is included.in this report by reference.
1 l
Amendment No. 25 to the license application dated September 6, 1963 (Docket No. 50-213) was submitted to the Commission effective March 11, 197h.
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This amendment consists of changes to the Facility Description and Safety Analysis and is included in the report by reference.
1 i
Amendment No. 26 to the license application dated September 6, 1963 (Docket No. 50-213) was submitted to the Commission on May 1h, 197h.
This amendment consists of Responses to Specific Questions I
asked by the Atomic Energy Commission concerning the Handling iJ of Spent Fuel Casks at Connecticut Yankee and is included in this report by reference.
l TECHNICAL SPECIFICATION CHANGES l
CY Change No. 2h (AEC Change No. 25)
Section 3.h (combined heatup, cooldown and pressure limitations) of the Technical Specifications with the associated curves, has been amended as a result of calculations, performed in accordance with Appendix G,Section III of the ASME Boiler and Pressure Vessel Code and Paragraph i
IS-522 of the Winter 1972 Addenda to ASME Section XI.
Test results of specimens from Surveillance Capsules "A" and "F" were involved in the establishment of revised operating curves for heatup, cooldown, hydrostatic
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and leak testing.
i CY Change No. 25 (AEC Change No. 26)
Section h.12 High Energy Piping System Tests This applies to inservice testing of the main steam piping system velds identified in Figure 4.12-1 whose failure could cause damage to the control room or switchgear room.
The objective is to provide assurance of the continued integrity l
of the main steam piping system over the plant's service lifetime.
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CY Change No. 26 (AEC Change No. 27)
Section h.ll Primary System Hydro Tests This applies to and specifies pressure test requirements of the primary system.
CY Change No. 27 (AEC Change No. 2h)
Section 3.17 Limiting Linear Heat Generation Rate This applies to measured peak linear generation cate (Kv/ft) in the reactor core for operating Cycle V and establishes limits which are based on the postulated loss of coolant accident (LOCA) vith appropriate allowances for fuel densification.
Section 3.18 Interim Limits for Power Distribution Control This interin specification applies to power distribution measurements, control rod group insertion and reactor thermal power levels for the first 6000 mwd /MTU of operation in Cycle V.
This specification vill be superseded at or before this exposure. The objective is to assure steady state operation within LOCA limits as specified in Section 3.17 and to minimize the potential for exceeding these limits during Xenon induced power peaking transients.
Part No. 5.0 Administrative Controls Section 5.3 Action to Be Taken in the Event of An Abnormal Occurrence in Plant Operation i
Revisions to Subparagraph "c" and "f".
Section 5.7 Plant Reporting Requirements Revisions to Subparagraph 5.7.1, 5.7.1.1 and 5 7.2. Addition of Subparagraph 5.7.1.J.
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'I PLANT OPERATING PROCEDURES The following new and revised procedures were issued during this report period.
NUMBER TITLE 11.51N2 Rev. 3 Reactor Containment Internal Filtration and Purging 2
ll.51N6 Rev. 1 Venting of H From Containment Following LOCA ll.51P1 Rev. 2 Waste Gas System: Purging, Normal Operations, Gas Discharge SOP #72 Temporary Reduction of Operating Pressure to 1900 psig RE-0P-9.0 Bank B Power Weighted Average Rod Height Log RE-0P-10.0 Average Bank B Position During Power Increase
" Hold Periods" RE-0P-11.0 Flux Map and Flux Trace Logging RE-0P-12. 0 Integration of Incore Flux Traces 1
RE-0P-15.0 Boration Requirements for Hot and Cold Shutdown RE-TP-6.0 Rev. 1 Rensselaer Polytechnic Institute - Nuclear Noise Measurement Test Special Maintenance Testing of 1250 Line Backup Tone Relaying Procedure Special Maintenance Replacing 21X Relay in 1250 Backup Relaying System Procedure l
RE-0P-5.0 Rev. 2 Fz Limit Calculations RE-OP-lh Core Power Tilt Determination 1
I QA-8.1 Identification of Materials, Parts, and Components QA-8.2 Material Issue QA-lh.1 Daily Plant Status Report AP 11 5h FAI.1 Flooding Alarm RHR l
AP 11.54 FAI.2 Flooding Alarm PAB East End j
AP 11.5h FAI.3 Flooding Alarm PAB West End t
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l AP 11.5h FAI.h Flooding Alarm PAB Pipe Chase East AP 11.54 FAI 5 Flooding Alarm PAB Pipe Chase West AP 11.Sh FAI.6 Flooding Alarm Metering Pump AP 11.Sh FAI.7 Flooding Alarm "A" Charging Pump AP ll.Sh FA1.8 Flooding Alarm "B" Charging Pump AP 11 54 FAI.9 Flooding Alarm Safety Injection Cubicle AP 11.5h FAI.10 Flooding Alarm Condene;te Return Tank PAB AP 11.5h FAI.11 Flooding Alarm Drumming Room AP 11.54 FAI.12 Flooding Alarm Screen Well House AP 11.5h FAI.13 Flooding Alarm EG-2A AP ll.5h FAI.lh Flooding Alarn EG-2B RP ll.51Q3 Rev. 2 Nuclear Instrumentation Operators Checkoff RP 11 51 P2D Original Test Tank Liquid Release Checkoff I
TIP 30 Original Change-over From Temporary PAB Ventilation System, Temporary Containment Purge System and Temporary Process Lines to the Newly Installed Systems TIP 29 Original Tie In Connections to Primary Drains Pump Discharge Line T-13, T-35, T-37, Flow Orifice Differential Taps and Installation of 2.127 Orifice Plate TIP 31 Original Tie In Connections to Boron Recovery System Vent Line To Waste Gas Header (T-78 and T-10)
TIP 32 Original Tie In T-122 Primary Water Supply to Primary Water Header in Radvaste Building LEM 3.2-7 Irradiated Fuel Transportation Incident LEM 3.2-15 12% Sodium Hypochlorite Spill EPL 1.5-1 Assessing Magnitude of Radioactive Iodine Release ADM 1.1-1 Conn. Yankee Organization Overall Responsibility and Authority ADM 1.1-2 Plant Operating Procedures QA 1.2-5.2 Procedure Format h
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QA 1.2-6.1 Document Distribution and Accountability QA 1.2-6.2 Master Document Index QA 1.2-6.h Temporary Procedure Changes QA 1.2-6.5 Procedure Review and Approval QA 1.2-18.1 Internal Audits 11.53 A7 Rev. 7 Spent Fuel Shipping with IF-200 Series Cask QA 1.2-16.1 Plant Information Report SUR 5.h-9 Accountability of Radioactive Material Released As A Result Of A Primary To Secondary Leak SUR 5.h-10 Accountability of Radioactive Gas Released to Stack When Sampling Reactor Coolant SUR 5.6 h Radiation Monitoring System MW-2P Liquid Monitor Calibration TIP 33 Tie-In Connection to Roof Drain (T-65)
TIP 3h Tie-In Connection for P-12h Discharge, New Diked Area Sump Pump, to Existing Drain System (T-64)
TIP 35 Tie-In Connection To The Existing Recycled Primary Water Makeup Line To The VCT Makeup Station T-2 11.51Q1 Rev. 4 Radiation Monitoring Operator's Check Off "7.
ABNORMAL OCCURRENCES i
NO.
DESCRIPTION Th-1 While conducting routine surveillance tests of the core cooling equipment initiation timers, No. 3 containment recirculating fan timer failed to actuate within the required time. The criteria for this timer is 4315 seconds; however, the timer actuated at 31 5 seconds.
The timer also failed to operate properly on two subsequent tests.
Since these timers function to sequentially load the emergency diesel generators during a simultaneous loss of AC power and core cooling actuation, operation outside the permissible time band is considered to constitute an abnormal occurrence.
Redundant backup timers were satisfactorily tested to insure proper response capability and the affected timer was replaced.
However, the replacement timer also failed to operate properly and further investigation revealed the problem to be malfunction of an SV relay located in the timer circuitry. The SV relay monitors h80V Bus 6 voltage and actuates the timer for the contain-cent recirculating fan provided the necessary core cooling signal is present. The SV relay appeared to be sticking and would not u
operate consistently.
The affected SV relay was replaced and successfully tested several times. The Plant Operations Review Committee investigated the incident and recommended an investigation to see if any other event is causing the relay failure. The committee also recommended a study to determine if SV relays need to be upgraded with a more reliable device.
Previous difficulty was experienced with the SV type relays in the timer circuitry during testing in October 1973, and was reported as Abnormal Occurrence 73-9 Additionally, during the refueling interval Loss of AC/ Core Cooling Test conducted on 12/1/73, SV relay 59-6 did not respond normally and was replaced with a relay of cimilar function. After each of these problems the affected relays and their associated timers were successfully tested and returned to service.
Presently the equipment timer surveillance test and the SV relay continuity checks are being conducted on a weekly basis to a
insure satisfactory operation. With weekly tests, plus redundant backup equipment it is felt that the system capability is adequate.
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Th-2 At 0612 on 1/18/Th. a data logger alarm message was received indicating a main steam line high steam flow condition in the
- 2 steam line. Two minutes later a similar high steam flow 1
condition was annunciated for the #1 steam line. The coincidence
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of two high steam flow signals satisfied the tripping logic and a reactor / turbine trip occurred from full power (N600 MWe).
Subsequent investigation revealed frozen instrument sensing l
lines leading to the differential pressure transmitters for the
- 1 and #2 main steam line excessive flow instruments. The instrument sensing lines were thawed and the instrumentation returned to normal.
The instruments in question are housed in an enclosure attached to the ou+nide of the reactor containment. Sub-zero tempera-tures sud strong winds created the freezing environment.
Additio tally, two unit heaters in the affected area were not operating.
The two unit heaters were repaired and returned to service. Two additional electrically operated, thermostatically controlled 2
heaters were placed in the affected enclosure for backup freeze protection. Several openings to the outside environment are in 1
the process of being sealed.
With the installation of backup electrical heaters and sealing the several openings, it is not expected that this situation will recur.
i Th-3 Station service power for the Connecticut Yankee Plant is supplied over two transmission lines which tie the Connecticut Light & Power Company system at the Montville generating station and Haddam substation (line 1206) with the Hartford Electric Light l
Company system at the Middletown generating station (line 772).
During a severe ice storm on January 19, 197h a momentary fault on the 772 transmission line caused an inadvertent trip of the j
1206 transmission line, resulting in a total loss of station service power. 'All systems responded normally to shut the plant 4
down to the hot standby condition. Both emergency diesel generators started and energized the respective emergency busses as required,
,l however, the service water pumps associated with the A emergency
'i diesel generator did not start automatically as required. The service ~ vater pump associated with the A emergency diesel was subsequently started manually. The auto-start capability of the service water pumps associated with the A emergency diesel generator was subsequently demonstrated several times.
The Middletown/ Connecticut Yankee transmission line outage was traced to a faulted lightning arrestor on the Pratt and Whitney
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Aircraft-Middletown (1572) transmission line.
Inadvertent tripping of the Montville/Haddam 1206 transmission line was apparently caused by incorrect blocking relay action, however, the exact cause is still under investigation.
The emergency generating system is equipped with undervoltage devices which strip and lock out all unnecessary 480 vctt loads to allov: diesel generator starting without load. Each h80 volt
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l bus has two Westinghouse type CV undervoltage relay circuits which trip and block auto closure of several components.
Redundant tripping capability is provided by undervoltage devices installed on the air circuit breakers of individual h80 volt components. Both undervoltage devices are equipped with time delay features to allow for 115 KV line reclosure or transfer of supplies.
It appears that the undervoltage device on the service water pump breaker actuated before the 480 volt bus undervoltage device, blocking the auto-start capability of 1A service water j
pump. The time delay features of both undervoltage devices were d
readjusted such that the Westinghouse CV type relay vill always j'
actuate before the undervoltage device on the service water pump 1
breaker. An investigation is being conducted to determine the f
necessity and desirability of removing the redundant undervoltage tripping device.
With readjustment of the time delay settings on the undervoltage l
devices, it is not expected that this situation vill recur.
Th-h The Connecticut Yankee Reactor Containment Atmospheric Control System includes a filtration system for removing radioactive
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fission products from the containment atmosphere following an incident which releases radioactive fission products from the core.
The filtration system utilizes the air recirculation fans as a driving force. The filters are located in four separate banks connected to the inlet plenums of the four air recirculation j
fans. The filter banks consist of moisture separators, absolute
. type filters and activated charcoal filters. Backup protection against overheating of the charcoal is provided in the form of a charcoal spray system. This system is included in each charcoal filter bank and supplies borated water from the Residual Heat Removal System to the face of the filters through nozzles.
During the routine weekly reactor containment inspection conducted on March 1, 1974, dry boric acid was observed on the floor adjaceent to the #h air recirculation fan unit. Further inspection within the filtration unit revealed boric acid deposits at the outlet or
. downstream side of at least twenty nine of eighty charcoal filter units. A detailed inspection of the charcoal filters from the inlet or upstream side revealed boric acid deposits of varying amounts on at least fifty five of the eighty filters in the
- h containment air recirculation fan unit. Leakage past the charcoal filter spray isolation valves provided the borated water path from the residual heat removal system to the filters in the
- h air recirculation unit.
Replacement charcoal filters have been ordered and are expected to be available within two to three weeks. Representative charcoal filters vill be removed from the #h air recirculation unit and will be tested for iodine removal efficiency. Those filters selected for efficiency testing vill represent the range from non-contaminated filters to filters having considerable boric acid deposition. The charcoal filter spray header isolation M "
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valve was disassembled and a mechanical obstruction in the gate valve groove was removed by filing.
The Containment Air Recirculation / Filtration System is designed to perform its function with only three of the four operable. The other three units were inspected and only one of two hundred forty charcoal units was observed to have slight boric acid contamination at the outlet. Further inspections from the inlet side of the other three units are planned.
74-5 The Reactor Coolant System letdown line is provided with a remotely operated isolation valve to automatically isolate the Reactor Coolant System from the Chemical and Volume Control System. This valve closes automatically upon safety injection actuation signal.
After completing a reactor coolant system cooldown and depress-urization evolution, the control room operator attempted to close the Letdown Motor Operated Isolation Valve (MOV-200) by actuating the control switch at the Main Control Board. The main control board valve indication showed that the valve had not completed its closing cycle. A local check revealed the valve in the half shut position. An attempt to manually close the valve failed.
An investigation by maintenance personnel revealed that the packing gland was tightened to the extent that the valve would I
not function properly.
It appears that the valve packing dried in the cooldown process.
It is our opinion that the valve would have operated as required with the plant at normal operating temperature.
This stop valve is located close to the loop piping and closes upon actuation of core cooling.
It functions to protect the reactor coolant system from a letdown line break. The only conceivable condition in which the public safety would be compromised is the simultaneous rupture of the letdown line and malfunction of the valve in question. Since the letdown line was not breeched during the period when the letdown line isolation valve was not operating properly, we conclude that the public health and safety was not compromised as a result of this occurrence.
The valve packing gland was adjusted followed by a successful operability check in the cold condition. The valve operability will be verified upon reaching normal operating temperatures.
74-6 A routine periodic Containment Penetration Leak Rate Test of P-60 (Component Cooling Water Supply to Neutron Shield Tank Cooler) was conducted on h-h-Th.
Mathematical calculations relating liquid test leakage to equivalent pounds of air @ 40 psig were performed which determined the penetration leakage to be 2.715 lbs.
air / day @ h0 psig. As the maximum leak rate for all containment penetrations is equivalent to 1470 lbs. air / day @ 40 psig, the leak rate for this one penetration exceeded the total. allowable.
'~~
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-32
l atmosphere.
However, since the Volume Control Tank hydrogen makeup is supplied through this regulator, flow is normally into the Volume Control Tank and most of the leakage was therefore from the Hydrogen Supply Tank. During periods when the Volume Control Tank conditions were such that no makeup was required, activity from the Volume Control Tank gas space could "back-up" through the inlet line to the leaking diaphragm; however, a check valve is installed in the inlet line and would severely limit flow in the reverse direction and would account for the lack of activity during the investigation.
The release of activity was dependent on at least three factors, (1) volume control tank conditions which required no hydrogen in-flow, (2) the check valve in the supply line which would prevent reverse (or outward) from the VCT, and (3) the small leakage permitted by the diaphragm seal. The gas release was therefore intermittent, at infrequent intervals and extremely limited in quantity.
Although the specific time at which the diaphragm seal began to leak cannot be determined, no activity was present in the Volume Control Tank until the plant startup on April 21.
Furthermore, air activity surveys within the auxiliary building in the vicinity of the Volume Control Tank were conducted following plant startup with no activity indicated. Additionally, no plant monitoring system indicated any unusual activity, nor 1
did any difficulty exist in operation of the Volume C,ntrol System. The activity detected on April 26 could not be repeated subsequently and surveys thorughout the area showed normal activity.
It is concluded therefore, that although no activity could be found at any time other than the event reported on the 26th and the discovery of the leaking diaphragm on May 3, it is possible that the diaphragm leak began during startup operations prior to the 21st; even so, the release of activity could only have been intermittent, at times when the Volume Control System did not require make-up and any such release vould be severely restricted and perhaps prevented in most cases by the check valve in the Volume Control Tank inlet line.
The isotopes released and their respective concentrations at the nearest site boundary are tabulated below:
Xc-133 3.73 x 10-11 uc/cc Xe-135 3.24 x 10-11 pc/cc Xe-135m 1.06 x 10-11 pc/cc Kr-87 3.7 x 10-12 pefce Kr-88 1.26 x 10-1I pc/cc Rb-88 9 35 x 10-13 pc/cc fio credit is taken for the decay of the released isotopes, from time of release to the time they reach the nearest
[ -
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site boundary. A conservative atmospheric diffusion factor was utilized in the calculation of activity at the nesrest site boundary.
l The Hydrogen Regulator has been replaced and it is not expected that this condition vill reoccur. Additional monitoring and leak checking is also planned in this area.
Th-8 Connected to the h.16 KV Emergency Diesel Generator Main leads is a 3 phase Auxiliary Transformer rated h5 KVA total, used to provide a 2h0V AC input to the Main Generator Field Circuit.
This Auxiliary Transformer is comprised of 3 single phase transformers, each rated 2h00/2h0V 15 KVA, connected in W e y
on high voltage side and Delta on the low voltage side.
During a routine monthly load test of 2B Emergency Diesel Generator, smoke was observed coming from the associated remote excitation cabinet. The rear panel of the excitation cabinet was removed and one of the three field excitation transformers was found smoldering. The unit was removed from s,ervice and the smoldering transformer was extinguished.
An inspection revealed a hole in the high voltage winding.
Transformer viring suffered considerable damage and the transformer itself was badly charred. A second exciter transformer in the same cabinet also exhibited evidence of overheating.
An improper grounding scheme caused a circulating current, resulting in an over loaded transformer and subsequent failure.
Failure of the exciter power transformer for Emergency Diesel Generator 2B did not degrade the emergency generation capability since a second completely redundant Dmergency Diesel Generator was available and operable.
During the outage of 2B Dmergency Diesel Generator, 2A Emergency Diesel Generator was started daily and all equipment associated with it was operable in accordance with technical specification requirements.
Two of the three excitation transformers and associated viring were replaced. After a series of pre-operational checks the
}
2B Diesel Generator was started, phased and loaded up to 2850 KW.
Observed line current readings indicated that the excitation power transformer was heavily overloaded. The readings were considered to be indicative of a circulating current in the lov side Delta. The vendor's representative recommended that the Auxiliary Transformer or grounding connection be lifted and the line currents measured once again. Lifting this Auxiliary Transformer ground connection resulted in normal transformer current measurements.
It has been concluded that removing this grounding connection is a satisfactory interim solution, since it reduced the loading of the Auxiliary Trans-former to a satisfactory level. We are presently investigating an improved ground scheme that should prevent the flow of
~
~ -
3
l circulating current.
Since Emergency Diesel Generator 2A has similar excitation transformer equipment and a similar ground scheme, a similar test was performed on this unit. The Diesel Generator was started, phased and loaded to 2850 KV and line currents were measured. A similar overloaded condition was observed. The current readings returned to normal after the grounding connection was removed.
With the installation of an improved grounding scheme, it is not expected that this situation will reoccur.
Th-9 During a routine bi-monthly Nuclear Overpower Trip Setpoint Check, two of four Nuclear Overpower Trip Setpoints were found 1.0% over the required 109%. The cause was determined to be normal instrumentation drift and trip setpoints too close to Technical Specifications limit. The Nuclear Over-power Trip Setpoint data over the past four years was reviewed and the instrumentation setpoint drift averaged N 0.5%.
This is well within the expected range of 2%.
Reactor Neutron Flux at power is measured by four separate channels.
If the power level reaches 109% of full power on any of two of the four channels a reactor trip will be initiated.
The setpoint of 109% includes an allowance for drift and setpoint error, calorimetric error and flux deviation due to rod motion. The maximum overpower correction prior to reactor trip that could result from the worst possible combination of these errors is 118% of rated power. The reactivity accidents analyzed in the Facilities Description and Safety Analysis are based upon an overpower reactor trip at the 118% setting.
We therefore conclude that the 1% drift in setpoint to 110% on two channels was within the normal expected instrumentation drift range, as described in the FDSA.
The setpoints on the two channels were adjusted to N 107.5%
to allow for normal instrument drift without exceeding the specified limit of 109%.
It is not expected that this situation will reoccur with the Overpower Trip Setpoints readjusted to allow for normal instrumentation dri ft.
7h-10 In accordance with Section h.9 of the CY Technical Specifications, the Main Steam Isolation Valves are tested each week for movement of the Valve Disc through a distance of approximately one and one-half inches. Upon actuation of the test switch for the
- 2 Main Steam Isolation Valve, no valve movement was noted.
The other three main steam Isolation Valves were successfully tested. The packing gland was adjusted followed by a successful valve movement test. The test was repeated a seond time and the valve responded normally.
It appears that the valve packing had dried out, restricting valve movement.
~
~~
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e i
The malfunction of the #2 Main Steam Isolation Valve did not present a significant hazard to the safety of the public. Even if a steam line rupture occurred coincident with the malfunction of the Main Steam Isolation Valve, analyses have demonstrated that neither DNB nor fuel damage vill occur as e result of any of the steam line rupture incidents.
Early in the plants' lifetime an incompatability was discovered between the packing material and the valve stems of the Main l
Steam Isolation Valves, resulting in corrosion of the valve stem and subsequent binding of the stem. Design changes i
were made by selecting a different packing material and installing an on-line test. stroking device. Test stroking has been conducted regularly and a review of the test results indicates only three cases in nearly nine hundred valve strokes that revealed adnormalities. The three problems were corrected by adjusting the valve packing gland followed by normal operation of the valve. The valve in question is a 2h" Main Steam Trip Valve manufactured by Schutte and Koerting j
Company.
Th-ll In accordance with the surveillance requirements of Section h.2 of the Connecticut Yankee Technical Specifications, Pressurizer Pressure Instrument Setpoints are tested every six weeks. During the pressurizer pressure surveillance test conducted June 20, 197h, one of the three low pressurizer pressure trip units operated approximately 15 psi beyond the acceptance criteria.
A variable low pressure trip signal is generated whenever
{
the pressurizer pressure is lover than a calculated variable i
low pressure setpoint. This setpoint is continuously calculated from measurments of reactor coolant temperature and core power.
Each of the three independently calculated setpoints is continously compared with one of the three pressurizer pressure measurements, giving three independent trip channels. Any trip signal generated from these channels is transmitted to a matrix of relay contacts l
vhich initiates a reactor trip upon coincidence of two of the three signals. Two of the three trip channels were within e
acceptable limits, however the third channel was 15 psig outside acceptable limits. Upon discovery of the out of specification condition, a trip signal was inserted to satisfy the one out of. tvo logic required for continued full power operation.
It is our opinion that this slight drift outside specified limits did not present a hazard to the health and safety of the public.
The faulty trip unit was replaced and is presently under-going bench tests in an attempt to determine the cause of the incorrect trip setpoint. The new unit was tested and subsequently returned to' service. The trip setpoint was adjusted to allow for normal instrumentation drift. The trip / alarm unit is a Foxboro Model 63S. There is no record of previous problems with this equipment.-
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UNUSUAL EVENTS i
NO.
DESCRIPTION Th-1 During routine testing of the Steam Generator Seismic Holddown
+
l Bolt Alarm System, on April h, 1974, one of the long bolt alarm relays was discovered charred and closed mechanically so that the alarm relay could not drop out.
The relay was replaced and
(
retested successfully. The relay was charred due to inadequate ventilation within the alarm relay enclosure. Ventilation was improved by installing louvers in the relay enclosure door.
All of the relays used in the system are Allen Bradley type 700 DC-M30021.
With improved ventilation within the relay enclosure, it is not expected that this situation will reoccur.
CHANGES IN KEY PERSO:mEL Mr. Norman A. Burnett, Shift Supervisor has been promoted g
to Operating Supervisor to fill a vacancy created by the resignation of Mr. Paul F. Ahern. All other key supervisory positions remain unchanged.
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POWER GENERATTON 197h l j,' '
Total Total January February March April May June For Year To Date Gross Thermal i
Power Generated 1272859 1160382 961813 387823 1309656 1236073 6328606 75L86175 in MWtH Gross Electrical Power Generated 421043 383255 315265 126656 L27869 393950 2068038 2k8230h1 l
in MWeH Net Electric 400668.3 36hh8h.9 299903.6 120230.5 40701h.6 3739h5.7 19662h7.6 23575369 7 s,
l6 Power Generated y
in MWeH Number of hours Reactor was critical 735.7 672.0 588.0 h66.6 744.0 712.0 3918.3 h9865.2 Number of hours Generator on line 726.75 660.h3 Shh.28 233.82, Thb.0 l 711.27 3620.55 h6075.8 I
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CONNECTICIT1 YANKEE ATOMIC POWER COMPANY 1
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10 15 20 25 30
f PLANT PERFORMANCE PRIMARY January Reactor performance was normal throughout the month. Results of flux caps taken vere normal.
February Reactor performance was normal throughout the month. Results of the full core map taken on February 14, 197h were normal. The boron /
burnup plot averaged well with predictions.
March Reactor performance was normal while operating during the
}
month. Results of the Full Core Flux Map taken on 3-21-74 were normal.
The core was symetric with a slight peak to the bottom. Margin between ceasured and limiting Fqn was 9.9%. The boron /burnup plot agrees reasonably well with normalized predictions. A Fz limit calculation was rum on 3-10-7h @ 80% power and indicated a 6.Th% margin to the allowable limit.
April Reactor performance was normal while operating during the month. The final results of a full core flux map taken on h-26-Th have not been received but the traces compared favorably with normal except for a slight increase in existing " blips". A Fz margin calculation was performed on h-23-74 following a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> hold @ 80%
power. The margin was calculated to be 10.3%. The boron /burnup curve slope is running shallover than predicted, 110 ppm B/1000 MWtD/MTU as compared to the normalized prediction of 118 ppm.
I May l
Reactor performance was normal while operating during the month.
i The results of flux maps taken in April and on May 16 compared favorably with those expected. The May map indicated the slight blip for incore positions H-13 and K-13 were in the magnitude of =1.0%.
Batch #6 is still the limiting fuel batch with the Fqn margin increasing with burnup.
Effective full power hours exceeded 3000 on May 29 which has increased the KW/Pr limit from 12.9 to 13.5.
This increase provides a larger Fz margin than previously for return to full power following load transients and the required (Tech. Specs. 3.17) reduced load holding periods. Continued Core Surveillance and data acquisition vill determine any necessary changes in the limiting fuel batch.
~
- 5.
r-The following Fz calculations were performed during the month:
Date Reason
% Margin to Allowable 5-5-Th Before return to 100% power after 11.01 holding period at 95% power follow-ing turbine valve test.
5-12-Th Before return to 100% power after 11.37 holding peri'od at 80% power follow-ing Steam Generator Feed Pump _
motor maintenance.
5-17-74 Before calculated Fz following the 10 96 monthly flux map, i
5-26-Th Before return to 100% power after 10 96 holding period at 80% power follow-e ing turbine valve test.
t 4
f The slop'e of the boron /burnup curve at the end of May was 107 ppm boron /1000 MWtD/MTU burnup. This is approximately 0.h% shallower j
than the normalized prediction of 118 ppm boron /1000 MWtD/MTU burnup.
Hand calorimetrics taken during the month were normal for the existing plant conditions.
i Jue i
Reactor performa'nce was normal while operating during the month.
The monthly full power flux map was taken June 11 and the results are normal with peaking factors becoming lower as burnup increases. The axial and radial flux shapes are becoming flat as expected. Batch #6 E
remains the limiting fuel batch, with Fq margin increasing with burnup.
n The following Fz calculations were performed during the month:
j Date Reason
% Margin to Allowable 6-9-Th Before return to 100% power after 17.8 holding period at 95% power follow-ing turbine valve test.
j
~
6-11-Th During routine monthly flux map 17.7
.at 100% power.
6-16-Th Before return to 100% power after 17.8 holding period at 80%. power follow-ing feed pump repair.
i 6-26-Th Before return to 100% power after 18.1:
holding period.at 80% power follow-ing plant trip. ~
~
~
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The slope of the boron /burnup curve at the end of June was i
113.5 ppm boron /1000 MWtD/MTU. This value generally agrees with the l
I nonnalized prediction of 118 ppm boron /MTU. Utilizing this slope the estimated "0" boron concentration can be expected to occur during the middle of May 1975 SECONDARY' i
j January Secondary plant performance was normal throughout the month with one exception:
1.
Towards the end of the month condenser performance appeared to be decreasing.
i February i
The secondary plant performance was normal throughout the month with one exception:
1.
Condenser performance was poorer than expected with the
],
backpressure higher than normal for 35 F circulating water j
temperature. The condition is being investigated.
i:
-March The secondary plant performance was normal while operating during the conth with the following exceptions:
i 1.
Condenser performance was still poorer than expected with a higher than normal back pressure for the circulating water temperature. The condition is being investigated
=
and preliminary findings indicate some leakage through the i
Steam Dump System. Also contributing is a higher than l
normal terminal temperature differential (TTD) for "D"
reheater. The steam dump valves were re-lapped during the plant shutdown.
f l
2.
Higher than normal TTD for "D" reheater indicated pressure l
tube leakage and/or partial flooding of the tubes. Several i
leaking tubes in "D" reheater were identified and plugged
):
while the plant was shutdown during the latter part of March.
~
April i
Plant efficiency is = 2 to 3 MWe lower than expected which is attributed to the continued poor condenser performance (previously
(
' reported) and to the new low pressure turbine spindles. The terminal temperature differential (TTD) of #1B feedwater heater is 1 to 2 F higher than normal and will continue to be observed for any further changes. 'The performance of "D" moisture separator-reheater han
_~
M m'
jm.
Ci
- -h8--
}
I returned to normal following repairs during March.
t May Plant efficiency remains lower than expected contributing to a reduced MWe output for the plant. As previously reported this ir attributed to a poorer condenser performance than normally expected and to the new low pressure turbine spindles. Condenser back pressure remains higher than normn1 and investigation into the reasons is continuing.
June Plant efficiency is still lower than expected contributing to a reduced MWe output for the plant. As previously reported this is attributed to a poorer condenser performance than normally expected and to the new low pressure turbine spindles. The higher than normal back pressure appears to be connected with "A" condenser. Cleanliness factor calculations indicate that "A" is 13.75%, (879 90%) lower than the expected value of 93.65%. The investigation into the poor clean-liness factor for "A" condenser is continuing in an attempt to identify the exact cause(s). The TTD for 1B feedwater heater, reported in the April Report as being 1 to 2 F above normal has returned to normal for some unknown reason.
}
SURVEILLANCE All checks, calibrations and tests required during the reporting period have been completed.
Containment Penetrations tested during the report period with the test results are as follows:
Test Number Function Date Leak Rate f
(pounds air / day @ h0 psig)
P-34 Component Cooling Water From RCP Therman Barrier 3-31-74 1.6 P-81 Aux. Steam Generator Feedwater Supply 3-21-Th 25 P-61 Component Cooling Water From Neutron Shield Tank Cooler 3-31-7h o,00 P-38 Comp. Cooling Water to RCP Thermal Barrier 3-31-Th 15.77 P-63 Neutron Shi'ld Tank Fill Line 3-24-74 0.00 e
/
_hg_
a i
'l P-lh Vapor Seal Head Tank Drain 3-31-Th 0.19 l
P-12A Valve Stem Leakoff 3-31-74 0.28
)
P-80 Aux. Spray Line From Fire System 3-20-74 0.00 P-28 Component Cooling Water to RCP h-1-7h 0.00 P-29 Component Cooling Water from RCP Cooler h-1-Th 5.40 P-23A Containment Open Bulb System h-1-Th 1 31 P-23B Containment Closed Bulb System h-1-Th 0.56 P-3 Safety Injection h-1-74 79.08 P-13 Containment Sump Pump Discharge h-1-7h 0.0h P-33 Berueling cavity Purification h-1-Th 0.00 P-70 Instrument Air Supply h-2-Th 0.00 s
]
P-66 Ccmponent Cooling Water to Drain Cooler h-2-Th 0.00 P-62 Service Air to Containment h-2-Th 0.05 P-71 Primary Vent Header h-3-Th 2.33 P-60 Component Cooling Water Supply to Neutron Shield Tank Cooler h-5-Th 3.53 P-C Equipment Hatch h-5-Th 0.00 P-10 Reactor Coolant Letdown h-6-7h 1.39 P-11 Liquid Sample Lines h-9-74 0.92 9
P-h Pressurizer Relier Tank Vent h-ll-Th 0.07 P-15 P-16 Steam Generator Blowdown h-15-Th 10.29 P-17 P-18 P-A Personnel Hatch 5-15-Th 21.27 me m
gh-p
P-67 Component Cooling Water Primary 5-17-7h 17 98 Drains Cooler Return P-12B Neutron Shield Tank Sample Line 5-17-7h 0.33 P-22 Space Heating Condensate Return 5-20-74 9 15 P-B Electrical Penetration 5-21-7h 17.3h P-hD Containment Purge Air Supply 5-22-74 0.00 P-30 Space Heating Steam Supply 5-2h-Th 1.52 P-64 Air Monitor Sample from Containment 6-1-74 3.65 P-65 Air Monitor Sample to Containment 6-1-7h 3.81 P-23D Air Monitor Purge 6-1-Th 3.08 P-2h Safety Injection Recirculation Lines 6-5-7h 0.00 P-68 Primary Water to Containment 6-11-7h 0.00 P-23C Dead Weight Tester 6-11-Th 0.08 P-69 Loop Full 6-lh-Th 1.70 P-bl Loop Drain Header 6-lh-Th 2.Th P-D Dome Vent Flange (Top) 6-24-74 0.00 P-E Dome Penetration Flange (Side) 6-24-74 0.00 Total for the current surveillance period 208.53 Total Allowable (Tech. Specs. h.4) lh70.00 Penetration testing for surveillance period 6 is complete.
Penetration P-50, Fuel Transfer Tube, will not be tested during j
surveillance period 6 since a refueling is required.
wa
^
51_
]sk
karm %W ma M e
wpAW hr-PRIMARY COOLANT CHD!ISTRY 1974 Gross Suspended Iodine Dissolved Activity Solids Tritium 131I 131I/133I Hydro.
Li.
Boron Oxy.
C1 pH (pci/ml)
(ppm)
(pCi/ml)
(pCi/ml)
Ratio (cc/kg) (ppm)
(ppm)
(ppb)
(ppm)
@ 25 C Max.
6.02(-1) 0.32 5.08 8.5(-3) 1.82 35 1.98
.1400 5.0
<0.07 6.25 JAd.
Ave.
4.63(-1) 0.289 4.13 5 1h(-3) 1.20 26.5 1.50 272 2.27
<0 07 5 9h Min.
7.99(-2) 0.132 3.26 2 99(-3) 0.95 21 1.05 1211 0.0
<0.07 55 Max.
6.86(-1) 0.348 5.39 5 16(-3) 0 92 31.1 1.65 1205 0.0
<0.07 6.2 FEB.
- Ave, b.52(-1) 0.277 4.88 2 99(-3) 0.86 28.1 1.03 1166 0.0
<0.07 5.92 o
Min.
3.77(-1)
'O.230 h.37 1.75(-3) 0.80 26.0 1.00 1119 0.0
<0.07 5.h Max.
5.51(-1) 0.1156 7.23 9 23(-3) 0.890 32.0 1.h95
,h -
1830 0.0
<0.07 6.h0 4
] MAR.
Ave.
3.73(-1)
.0.0k69 7.18 5.89(-3)
O.837 28.0 1.254 1303 0.0
<0.07 6.3h Min.
1.3(-2) 0.0260 7 12 3.59(-3) 0.800 25.0 0.950 1070 0.0
<0.07 5.75 3.97 6.16(-3) 0.93 32 1.78 1833 0.0
<0.07 6.20 Max.
5.60(-1) 0.049 APRIL Ave.
1.k9(-1) 0.0h7 3 97 3.78(-3) 0.77 32 1.49 1522 0.0
<0.07 5.68 Min.
1.92(-h) 0.0hh 3 97 2.18(-3) 0.50 32 0.95 10h2 0.0
,g,g7 5,go Max.
6.20(-1) 0.047 6.40 h.h6(-3) 0 938 33.0 1.63 10h3 50
<0.07 6.h 6 gky Ave.
4.h9(-1) 0.039 6.33 4.30(-3) 0.906 29.4 1.24 loo 7 3.o
<0.07 6.2 l
Min.
2.10(-1) 0.030 6.27 4.0h(-3) 0.856 24.0 0.70 955 0.0
<0.07 6.1 Max.
5.75(-1) 0.131 8.9h 1.42(-2) 4.29 34.0 1.67 1012 0.0
<0.07 6.50 JUNE Ave.
h.95(-1) 0.127 8.75 6.h5(-3) 1.78 29 2 1.37 908 0.0
<0.07 6.27 Min.
h.15(-1) 0.123 8.39 3.36(-3) 0.91 24.0 1.07 8h9 0.0
<0.07 6.05
l-i t
t l SOLID WASTE SHIPMENTS 4
4 'j.
Description Activity Date of Item-Volume F3 in Curies Disposition Destination 4
2-19-74 125-55 Gallon 918.75 0.28h Offsite West Valley, j
drums compacted burial N.Y.
vaste 4
)
3-7-7h Spent IX Resin 100.00 75.0 Offsite Moorehead, Ky.
burial i
3-26-74 Spent IX Resin 100.00 31.0 offsite Moorehead, Ky.
burial 4
h-ll-Th 86-55 Gallon 632.1 0.148 offsite k'est Valley, drums compacted burial N. Y.
j waste h-11-Th 2 Wooden Crates 200.00 0.008 Offsite West Valley, 4
l, scrap lumber burial N
Y-i h-25-Th 122-55 Gallon 896.70 0.356 offsite West valley, drums compacted burial N. Y.
j waste i
5-31-Th Spent IX Resin 100.00 395.00 Offsite Moorehead, Ky.
l burial I i i
TRANSFER OF SPECIAL FRICLEAR MATERIALS i!
j-The following shipments of spent fuel to the processing facility at Morris, Illinois were made during the month of May.
Date Shipment No.
Assembly No.
5-15-74 75-2-24 B-20 5-21-7h 76-2-25 B-h8 J
5-31-74 77-2-26 B-37 i l
~
-- -i -
~~.
y_
-~,.
53
" ' ~ ^
n
=
x
l OCCUPATIONAL PERSONNEL RADIATION EXPOSURE Incremental Personnel Exposure Exposure Increments Number of Personnel Less than 0.100 rem 1437 0.100 to 0.250 rem 96 0.250 to 0.500 ren 61 i
j 0.500 to 0.750 rem 18 0.750 to 1.000 rem 13 1.000 to 2.000 rem 30 2.000 to 3.000 rem 8
3.000 to 14.000 rem 1
14.000 to 5 000 rem 0
5.000 to 6.000 rem 0
Greater than 6.000 rem 0
Exposure >0.500 rem according to Duty Function 1
Routine Operations 9
Routine Maintenance 13 Routine Instrumentation 2
Routine Health Physics and Chemistry 14 Special Maintenance Containment Coolers 2
Repacking Pressurizer Spray Valves 2
Removal and Re-installation of Insulation on Piping in the Containment 7
Removal and cleaning of Charcoal Filters in Containment 2
Solidification of Waste Material 1
Spent Resin Liner Movement 1
New Construction Piping Tie-ina 26 Tracing Piping for Procedure Writing 1
1
_53
..__--____..m.m-__.
__ _... -.._..__ _ _.m
__.______i..m.
.._ - _ _ m _ __ _ _.. _
i r
L i.
REPORT OF RADIDACTIVE EFFLUENTS l
l Facility: Connecticut Yankee Atomte Power Coipany Docket: 50-211 I.
LIQUID RELEASES (3)
,s t'n t r a Ja9.
Feb.
Nr.
t Aor.
May June b Month Total v
i
- 1. Cross radioactivity (6.v(7' a) Total release Curtes 2.59(-1) 1.46(-1) 9 k (-1 ) it. 51(-1 ) - cf 11 1.08(-1) 1.50 E+00 b) Avg. concentration 5.L5(-9)!5.53(-91l7.85f-0) l L
1.72(-9) k.k1 E-09 released tCt/mi 3.9s(-91 2.46(-9)
L c) Pax. concentration released LCt/mi 1.95(-7) 4.57(-8) 1,0a(-7) 7.00(-8) 5.86(-T) 3.9h(-8) 5.86 E-07 i
- 2. Tritiurs I
i a) Total release Curtes 7.84(1) 9.66(1) 1.08(2) 1.54(?) i6.23(1) 1.75(2) I 6.74 E+02 l
b) Avg. concentration C
released ret /-1 1.21(-6) 1.63(-6) 1.73(-6)5.6k(-6)j+o.78( 7 ) 2.79(-6) :
1.98 E-06
- 3. Dissolved noble gases I
s a) Totsi release Curtes 2.51(-2) 2.27(-2)
- 2. 8 3 (-2 ) i l.o (-2 ) :9.k3(-3) 1.22(-2)*
1 if t-01 I
g b) Avg. :oncentration i
released uct/ml 3.R 7(-!Oi 3. 8 2 (-101 h. 5 h (-10 '6.96(-10 ti.L8(-lo il.95(-10 3.kk E-10
~
[
- 4. Cross alpha radio-l
~
acttvity r
a) Total release Curteg I W E-U r
released LCt/mi g
i i
b) Avg. concentration l 2.1k E-11
- 5. Volume of liquid waste u
to discharee canal Liters 5.36(6) 5.31(6) 1.5k(6) 'k.73(6) l3.82f 6)5.07(6) 2.58 E+07 y
6.1*olus e of dilution water Liters 6.48(10) 5.94(10) 6.2h(10)c2.73 f10) 6.17(19) 6.27(10) : 3.k0 E+11 I
k'
- 7. Isotopes released "Curtes i
Kr-87 i
I-133 6
i i
I-131 i
i Xe-133 2.27(-2) 1.38(-2).1.8(-2) 1.15(-2) < Af 15 41.09(-?): 6.25 E-02 Xe-135 2.44(-3) 8.93(-3) 1.03(-2) 7.53(-3), air s p1.?c t_1) e 3.k3 E-02 Cs-137 9.00(-2)l).09(-2) 6.8(-?) 4 1_16(-3 )'7_ t,f 13 !s.0?r 1) i 1.83 E-01 r
Ce-144
- 4. 7s f-2)l 2.06(-2). t. L e f-? ) l l.7q t-? )it
,2 f _,i i 3 _1 o r -? ) I 1.90 E-01 Co-60_
4.72(-2)l 3.34(-2) l7. As(-?)( b 28 f-?) 3_ rt r 3 )t1.17(-?) i 3.66 E-01 Co-58 7.2 3(-?f 3,8 (-2) i6.17(-? ) i ?.96(-2 )' T. <n f 11 '! _0 A f -91 6 3.24 F-01
_Cr-51 2.23(-2) it.11(-1) is $4(-2)*1_ % f 1)i1_lo(-?)I k.03 E-01 Mn-54 t
8 r
- 1_7Lf 1)i5.7(-1) i 9.hk E-03 f
%-103 l
Sr-90 I
i i 8.01 E-04
{
sr-89 6
i i 8.60 E-OL
,e l'ntdentified (Cross S-v) 7.21(-4) 4. M ( /. ) s ? h (-h )
.s.g(-3) 6 9,L f _q i _qif _L ) i 7.58 E-03 i
I e
i e
i
- 8. Percent of Tech. Spec.
limit for total activity released (2) 6.7 (-3) 6.75(-3) 7.85(-3) 1.66(-2) 5.20(-3) 9.26(-3) 1.8~ E-01 (1) Except tritius and dissolves gases.
(2) Eased upen annual projected available dilution water.
(3) Differences between the su cf the monthly colu=n and the 6 =cuth colu=n comes frc= the results of co=posite se=ples analyses teir.g included in the 6 month total colu=n.
(k) Eased upon actual available dilution water.
t.
b
.____.m._
_.._._.____...___m
___.m__ _
_.___m__.
h ~
L-l 9
I II. AIRBORNE RELEASES (1)
Units Jai.
Feb.
I Mar.
Aor.
May June 6 Honth Total I. Total noble gases Curtes 1.28f-1) 1.08(-1)th 6hf-11 1.02 0.71(-91 1.57f 1) 1.97 E+00
- 2. Total halogens Curtes L_57( Al L.57 E-08 l
- 3. Total particulate i.
gross radioactivity (9.v) curies k.37(-8) 4.59(-7) 1 58( kl 1.59 E-Oh
- 4. Total tritiuts Curtes 8.61(-h)'
8.61 E-04
- 5. Total particulate l
gross alpha radio-activity Curtes
- 6. Eastuun noble gas
}
[
release rate uC1/see.il2.0 7.07 2.56(1) 31.6 11.6 7.89 3.16 E+01
- 7. Percent of applicable Itste fort a) Noble gases I
+2.81(-4) 2.5(-4) 3.68( k)lb.5H -4)t 2.1(-4 ) 3. 34 (-4 ). 1.92 E-03 i
b) Halocens I
IL.81(-8)I h.81 E-08 F
c) Particulates i
4 Ih.6f 8) 86.8hf-7) iS.5kf-7) 1.08 P ')6 d) Tritium I
i 8 h. M H -7 ) i L.51 F-07
- 8. Isotoce released:
- Curies 1 6
i 1
l Particulates f
I i
b
~
C-14 I
i 1.h9 E-03 Ba + 1.a-140 l
?
sr-90 t
t Cs-133 i
i 95-33 i
t
- 1.58( h) i 1.55 E-Ob t
Unidentif ted t
i k.37(-8)th.59f-7)'
5.03 E-07 Halonens i
1-131 4
k.57(-8)
L 57 F-08 l
1-133 i
[
1-135 i
Ga,es i
i Kr-45 6
- 2. M (-1 ) 7. 76 (-1 )
1.C6 E+00 I
xe-133 5.92(-2) 3.7(-21 ti.Ch (-1 )
9 itf 11 3,nt-9) 6.16(-2) 5.11 E-01 Kr-88 2.55(-2) 2.27(-2)12.L6(-2) 1.13(-2) 9.0!f 91 3.11(-2)-
1.35 E-01 r
Kr-87 7.73(-3) 7.21(-3)i8.5H-1) 1.54(-3) 7_nf-11 9 55(-3) 4.36 E-02 l
Kr-85m xe-13S xe-135,
[
t 1.43(-2) 1.45(-2) 1.16(-2) 3.8(-3) 7 oof-11 1.65(-2)I 7.06 E-02 xc-135 3.38(-2) ' 2. 69 (-2) '2.73(-2 ) 1.27(-2) 9.Lar 91 3.dk(-2) i 1.63 E-01 Ar-41 e
6 i k.50 E-03 Ar-37 1
4 i
r i
f I
i I
i l
I I
(1) Differences between the su. of the monthly colun and the 6 month colu:r.n comes fro:t the results of sample analyses perfor=ed off-site being included in the 6 month colunn.
s
RALIOLOGICAL ENVIRotNENTAL MONI'lDRING PROGRAM SU'CIARY OF DATA (1)
Number Highest Average Activity Number Total Locations Iocation Activity Medium of Number of Above Distance Sampled Farples Sarnles Ground (miles)
Direction Units High Ave.
Low Basi s Comments Air Gross 3
0.3h 0.18 0.0h Beta Fallout' levels Particulate 10 208 0
2.0 SE pCi/m Soil 10 10 0
3.5 bcal PC1/g L.3 L.3 L.3 Gross Background levels Gamma Vegetation 10 10 0
8.0 SSE pCi/g 1.2 1.2 1.2 Gross Background levels Ga=ma Well Water h
12 0
1.0 S
pCi/l 3.5 35 3.5 Gross Background levels Beta i
f 1
Milk 6
18 0
0C1/1
<0.5 <0.5
<0.5 I-L31 No Detectable activity Ga==a Gamma Direct radiation from Exposure lh 8h 3
0.5 N
pr/hr 25.2 26.h 20.6 Dose Waste Gas Sphere Botto=
Gross Plant Discharge g
Sediment 3
6 1
05 ESE pC1/g 7.7 5.6 3.5 Gamma Canal River Water h
8 0
0.5 SSW pCi/1 7.0 5.0 3.0 Gross
Background
Beta Variations Fish 3
8 0
0.5 SSW pCi/g 0.23 0.23 0.23 Gross Fallout levels Ga==a Shellfish 2
3 0
05 SSW pCi/g 0.32 0.16 <0.09 Gross Fallout levels Ga=me (1) Supporting information for this su==ary has been forwarded to the AEC under separate cover.